Candidate Materials For Fusion Reactors Research Articles

A Novel Low-Activation VCrFeTaxWx (x = 0.1, 0.2, 0.3, 0.4, and 1) High-Entropy Alloys with Excellent Heat-Softening Resistance.

The microstructure, Vickers hardness, and compressive properties of novel low-activation VCrFeTaxWx (x = 0.1, 0.2, 0.3, 0.4, and 1) high-entropy alloys (HEAs) were studied. The alloys were fabricated by vacuum-arc melting and the characteristics of these alloys were explored. The microstructures of all the alloys exhibited a typical morphology of dendritic and eutectic structures. The VCrFeTa0.1W0.1 and VCrFeTa0.2W0.2 alloys are essentially single phase, consisting of a disordered body-centered-cubic (BCC) phase, whereas the VCrFeTa0.2W0.2 alloy contains fine, nanoscale precipitates distributed in the BCC matrix. The lattice parameters and compositions of the identified phases were investigated. The alloys have Vickers hardness values ranging from 546 HV0.2 to 1135 HV0.2 with the x ranging from 0.1 to 1, respectively. The VCrFeTa0.1W0.1 and VCrFeTa0.2W0.2 alloys exhibit compressive yield strengths of 1341 MPa and 1742 MPa, with compressive plastic strains of 42.2% and 35.7%, respectively. VCrFeTa0.1W0.1 and VCrFeTa0.2W0.2 alloys have excellent hardness after annealing for 25 h at 600–1000 °C, and presented compressive yield strength exceeding 1000 MPa with excellent heat-softening resistance at 600–800 °C. By applying the HEA criteria, Ta and W additions into the VCrFeTaW are proposed as a family of candidate materials for fusion reactors and high-temperature structural applications.

The plasma focus as a tool for plasma-wall-interaction studies

The study of the interaction of magnetized plasmas with candidate materials for fusion reactors, as for example tungsten, is a main topic in fusion research. Many studies simulate the plasma wall interaction using ion beams, while only a few use plasma simulators. Plasma foci can produce dense magnetized plasmas of deuterium and helium among other species. We used the plasma focus Fuego-Nuevo II, to expose tungsten samples to deuterium and helium plasmas. The samples were analysed by means of SEM, RBS and NRA, evidencing surface erosion, surface melting and retention of deuterium in a shallow surface layer of 250 nm amounting 6.5·1016 D/cm2. The plasma temperature has been measured at the position of the samples using a triple Langmuir probe and compared to calculations of a snowplow model. The modelling of the electrode to reach desired plasma parameters is discussed.

The European contribution to the development and validation activities for the design of IFMIF lithium facility

The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where candidate materials for fusion reactors will be tested and validated. The high energy neutron flux is produced by means of two deuteron beams (total current of 250mA, energy of 40MeV) that strikes a liquid lithium target circulating in a lithium loop of IFMIF plant. The European (EU) contribution to the development of the lithium facility comprises five procurement packages, as follow: (1) participation to the experimental activities of the EVEDA lithium test loop in Oarai (Japan); (2) study aimed at evaluating the corrosion and erosion phenomena, promoted by lithium, for structural fusion reference materials like AISI 316L and Eurofer; (3) design and validation of the lithium purification method with the aim to provide input data for the design of the purification system of IFIMF lithium loop; (4) design and validation of the remote handling (RH) procedures for the refurbishment/replacement of the EU concept of IFMIF target assembly including the design of the remote handling tools; (5) the engineering design of the European target assembly for IFMIF and the safety and RAMI analyses for the entire IFMIF lithium facility.The paper gives an overview of the status of the activities and of the main outcomes achieved so far.

Study of an experimental system for hydrogen permeation in thin-solid materials for fusion plasma research

Tritium permeation is one of the key issues in fusion reactor operation. However, detailed permeation data for hydrogen and its isotopes in the fusion reactor candidate materials are very limited. In order to investigate the permeation behaviors of hydrogen in the thin-solid materials, we constructed an experimental system adopting a continuous flow method; high vacuum of 10 −9 Torr and sample heating upto 1273 K were successfully achieved. Hydrogen permeation experiments were also performed by using thin-solid samples of stainless steel (SS 316L); the results were compared with the previously existing reference data.

Analysis of displacement damage in materials in nuclear fusion facilities (DEMO, IFMIF and TechnoFusion)

Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al 2O 3, SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.

Thermo-structural analysis of target assembly and back plate in the IFMIF/EVEDA lithium test loop

IFMIF is an accelerator-based intense neutron source for testing candidate materials for fusion reactors. The Engineering Validation and Engineering Design Activities (EVEDA) of IFMIF are in progress under the Broader Approach Agreement. In the EVEDA Li test loop, two types of target assemblies, “an integrated type target assembly without a back plate”, made of SUS316L, and “a target assembly with replaceable bayonet back plate”, made of F82H steel, will be evaluated. For these target assemblies, thermal stresses were evaluated using the ABAQUS code. As the permitted stress, the values of yield strengths at 300°C were used. In the case of the integrated target assembly, the maximum thermal stress was below the allowed stress (164MPa). However, in the case of the target assembly with the bayonet back plate, thermal insulation around the back plate should be necessary to reduce the thermal stress below the allowed value (455MPa).

Performance analysis of fusion nuclear-data benchmark experiments for light to heavy materials in MeV energy region with a neutron spectrum shifter

Nuclear data are indispensable for development of fusion reactor candidate materials. However, benchmarking of the nuclear data in MeV energy region is not yet adequate. In the present study, benchmark performance in the MeV energy region was investigated theoretically for experiments by using a 14MeV neutron source. We carried out a systematical analysis for light to heavy materials. As a result, the benchmark performance for the neutron spectrum was confirmed to be acceptable, while for gamma-rays it was not sufficiently accurate. Consequently, a spectrum shifter has to be applied. Beryllium had the best performance as a shifter. Moreover, a preliminary examination of whether it is really acceptable that only the spectrum before the last collision is considered in the benchmark performance analysis. It was pointed out that not only the last collision but also earlier collisions should be considered equally in the benchmark performance analysis.

Measurement of Wavy Surface Oscillations on Liquid Metal Lithium Jet for IFMIF Target

Abstract The International Fusion Materials Irradiation Facility (IFMIF) has been conceived as a high-flux 14 MeV neutron source for testing candidate fusion reactor materials. In the current design, neutrons are generated by irradiating a target with a deuteron beam and high-speed free-surface flow of liquid metal lithium (Li) is adopted as the target. To reveal the stability of the Li flow, we have examined characteristics of surface waves at a location 175 nm downstream from a nozzle exit, which corresponds to the center of the beam irradiated region. In this study, the characteristics of surface waves just downstream of the nozzle exit were measured experimentally, since the initial growth of surface waves exerts a definite influence on the surface behavior of the Li flow in the downstream region. Experiments were carried out with a focus on surface oscillations of the Li flow using the lithium circulation loop at Osaka University. These oscillations are measured using an electro-contact probe apparatus, which can detect electrically a contact between the probe tip and the Li surface and provide local height data of surface waves. The apparatus was installed at a location 15 mm downstream from the nozzle exit and scanned the Li surface by moving along the liquid-depth direction. The experiments were performed for the velocity range of 3-15 m/s under argon gas atmosphere at a pressure of 0.13 MPa. The contact signal recorded in the experiment was used to analyze the characteristics of surface waves, and then the root-mean-square wave amplitude and the frequency of surface waves were calculated. It was found that the root-mean-square wave amplitudes of surface waves increased with a rise in the flow velocity, and reached approximately 0.18 mm at 14-15 m/s. And also, obtained frequencies were analyzed using a linear stability theory, and the variation of frequencies was examined with the mean flow velocity.

Design and Construction of IFMIF/EVEDA Lithium Test Loop

Abstract The International Fusion Materials Irradiation Facility (IFMIF) is a D+–Li neutron source aimed at producing an intense high energy neutron flux (2 MW/m2) for testing candidate fusion reactor materials. Under Broader Approach activities, Engineering Validation and Engineering Design Activities (EVEDAs) of IFMIF started on July 2007. Regarding the lithium (Li) target facility, design, construction, and tests of EVEDA Li test loop (ELTL) is a major Japanese activity. The detail design of the loop was started since early 2009. Construction of the loop was started at the middle of 2009, and completion is scheduled at the end of February 2011. This paper focuses on the design of the loop configuration and the major components. ELTL was designed to consist of two major Li loops, which are a main loop and a purification loop including an impurity monitoring loop. The main loop equips a target assembly which produces a high-speed free-surface Li flow to test the flow stability as the D+ beam target. The maximum flow rate of an electromagnetic pump in the main loop was set to 3000 l/min, so that flow velocity in the target assembly is 20 m/s at the maximum. Regarding the purification loop, a cold trap and two hot traps and impurity monitors are installed in order to purify and monitor impurities in Li. The configuration of these components in addition to the specification and configuration of the whole loop is presented.

IFMIF High Flux Test Module—Recent progress in design and manufacturing

The International Fusion Material Irradiation Facility (IFMIF) is an accelerator driven neutron source for irradiation tests of candidate fusion reactor materials. Within the High Flux Test Module (HFTM) a testing volume of 0.5 l filled by qualified small scale specimens will be irradiated at displacement rates of 20–50 dpa/fpy in structural materials. The Forschungszentrum Karlsruhe (FZK) has developed a HFTM design which closely follows the design premise of maximising the space available for irradiation specimens in the IFMIF high flux zone and in addition allows keeping the temperature nearly constant in the rigs containing the specimen. Complementary analyses on nuclear, thermo-hydraulics and mechanical performance of the HFTM were performed to optimize the design. The present paper highlights the main design characteristics as well as recent progress achieved in this area. The contribution also includes (i) recommendations for the use of container, rig and capsule materials, and (ii) a description of the fabrication routes for the entire HFTM including brazing and filling procedures which are currently under development at the Forschungszentrum Karlsruhe.

Design of a lip seal-replaceable backwall for IFMIF liquid lithium target

IFMIF is an accelerator-based intense neutron source for testing candidate materials for fusion reactors. Neutrons are emitted inside liquid Li flowing on a so-called backplate or backwall made of reduced activation ferritic/martensitic (RAFM) steel. The backwall is joined to a target assembly by means of a lip seal welded by a remote handling system. Since the backwall is operating under severe neutron irradiation conditions, such as a damage rate of 50 dpa/year and a maximum nuclear heating rate of 25 W/cm 3, thermo-structural design is one of the critical issues for target design. To reduce thermal stress in the lip seal region, a stress mitigation structure was provided. Thermal stresses were calculated by using the ABAQUS code. The analysis of the lip seal with the stress mitigation structure showed that the maximum von Mises stresses of the backwall core (RAFM) and the lip seal (316L) were 93 and 335 MPa, respectively. Considering the existing weld technique, a modified-improved model to be realized by using the TIG and YAG welding is proposed. R&D on the lip seal weld is currently under way.

Thermo-structural analysis and design consideration of the replaceable backwall in IFMIF liquid lithium target

The IFMIF is an accelerator-based intense neutron source for testing candidate materials for fusion reactors. Intense neutrons are emitted inside the Li flow through a backwall. The backwall made of 316L stainless steel or RAFM is attached to the target assembly with a lip seal welded by a YAG laser. Since the backwall is operating under a severe neutron irradiation of 50 dpa/year and a maximum nuclear heating rate of 25 W/cm 3, thermo-structural design is one of critical issues in a target design. Thermal stress was calculated using the ABAQUS code. As a permissible stress, yield strength at 300 °C was used. In the case of the 316 stainless steel backwall, the maximum thermal stress was more than the permissible stress (164 MPa). On the other hand, in case of the F82H backwall, a maximum thermal stress was below the permissible stress (455 MPa). Therefore, F82H is recommended as the backwall material.

High temperature indentation tests on fusion reactor candidate materials

Flat-top cylinder indenter for mechanical characterization (FIMEC) is an indentation technique employing cylindrical punches with diameters ranging from 0.5 to 2 mm. The test gives pressure–penetration curves from which the yield stress can be determined. The FIMEC apparatus was developed to test materials in the temperature range from −180 to +200 °C. Recently, the heating system of FIMEC apparatus has been modified to operate up to 500 °C. So, in addition to providing yield stress over a more extended temperature range, it is possible to perform stress–relaxation tests at temperatures of great interest for several nuclear fusion reactor (NFR) alloys. Data on MANET-II, F82H mod., Eurofer-97, EM-10, AISI 316 L, Ti6Al4V and CuCrZr are presented and compared with those obtained by mechanical tests with standard methods.

Neutron imaging of fiber-reinforced materials

Glass-fiber-reinforced plastic laminates used for the insulation of Toroidal Field (TF) magnet-coils and fiber-reinforced silicon carbide ceramic composites used as structural material for the self-cooled Pb-17Li blanket module are attractive candidate materials for fusion reactors because of their high performance under extreme conditions. Porosity, which depends on the manufacturing process, and swelling of fiber-reinforced materials due to the high flux of radiation are the main problems. The aim of this study is to describe the experimental procedures of different imaging methods, and also to decide the most efficient imaging method for the investigations of the complex microstructure of fiber-reinforced materials. In this work, the fiber-reinforced composites were inspected with neutron and X-ray radiographies at ATI-Vienna and also at PSI-Villigen. A contrast enhancement at the edges can be achieved by means of phase contrast neutron radiography (NR), which is based on the wave properties of neutrons and arises from the neutron refraction (rather than attenuation). Elements having different refractive index within a sample cause a phase shift between coherent neutron waves. The degree of coherence can be determined by means of the coherence pattern caused by the sample, when a point source (pinhole) is used and the distance between source and sample is varied.

Point Defect Formation in V-4Cr-4Ti and F82H Irradiated with Fission and Fusion Neutrons

Vacancies and interstitial type dislocation loops of two fusion reactor candidate materials (V-4Cr-4Ti and F82H) after fission and fusion neutron irradiation were studied by positron annihilation lifetime spectroscopy. Fusion neutron irradiation was performed in the FNS of JAERI, and fission neutron irradiation was performed in the KUR of Kyoto University. The neutron irradiation dose was about 1×10−6–1×10−4 dpa, and the irradiation temperatures were room temperature and 673 K. In the irradiation at room temperature, the defects (mainly dislocation loops) in both alloys were detected even at a low irradiation dose of 10−6 dpa and the mean lifetime of positrons increased as the irradiation dose increased. The effects of the fission and fusion neutron irradiation on the point defect production were almost the same if they were compared at the same dpa. This can be explained by the fact that the number of subcascades, which is an important factor for the defect formation at room temperature, is proportional to dpa in these metals. In contrast, an effect of cascade size was found for the irradiation at 673 K. Dislocation loops were detected only in the fusion-neutron-irradiated F82H at 673 K.

Effect of helium on dislocation loop formation and radiation swelling in SiC

SiC and SiC/SiC composites are considered as candidate materials for fusion reactors. In a fusion reactor environment, helium atoms will be produced in SiC up to very high concentrations (15 000–20 000 at. ppm) and therefore it is very important to understand how helium effects radiation swelling of SiC. In this paper a theoretical model of the helium effect on radiation swelling of SiC is suggested. This model is based on considering of kinetic growth of dislocation loops in the matrix taking into account the effect an internal electric field formed near dislocation loops has on diffusion processes of charged point defects. The trapping of helium atoms by vacancies results in an enhanced growth rate of dislocation loops and finally a swelling increase. The theoretical results for radiation swelling are compared with the existing experimental data. It is shown that helium atoms increase the radiation swelling of SiC, especially at high temperatures.

Effect of Helium on Radiation Swelling of SiC

Ceramic materials produced on the basis of SiC and SiC/SiC composites are considered, due to their high-temperature strength, pseudo-ductile fracture behavior and low-induced radioactivity, as candidate materials for fusion reactors. The radiation resistance of ceramic materials under neutron irradiation is one of the key problems, which determines the use of these materials in fusion reactor environment. The dimensional stability (radiation swelling) of SiC is an important problem for degradation of radiation resistance of SiC materials. In fusion reactor environment helium atoms will be produced in SiC in the first wall region up to very high concentrations (15,000–20,000 at.ppm) and therefore it is very important to understand the helium effect on radiation swelling of SiC.In the present paper the effect of helium on radiation swelling of SiC is investigated. Recent experimental results concerning the helium effect on radiation swelling of SiC under neutron, single-ion and dual-beam irradiation are presented. A new theoretical model is suggested for an explanation of the effect of helium on radiation swelling in ceramic materials. Point defects in ceramic materials can have an effective charge (e.g., an F+centre, vacancy with a single trapped electron). The suggested theoretical model is based on kinetic consideration of charged point defect accumulation and kinetic growth of dislocation loops in a matrix of ceramic material (SiC) taking into account the effect of internal electric fields formed under irradiation near dislocation loops on diffusion processes of charged point defects. In this theoretical model, vacancies and small vacancy clusters are considered as additional traps for helium atoms. This results in enhanced growth rate of dislocation loops and finally swelling increase. The obtained numerical and analytical theoretical results for radiation swelling in the presence of helium atoms are compared with the existing experimental data for irradiated SiC material containing helium atoms. It has been shown that helium atoms increase the radiation swelling of SiC.

Radiation swelling of SiC under neutron irradiation

Ceramic materials produced on the basis of SiC and SiC/SiC composites are considered due to their high temperature strength, pseudo-ductile fracture behavior and low-induced radioactivity as candidate materials for fusion reactors. The radiation resistance of ceramic materials under neutron irradiation is the key problem which determines the use of these materials in fusion reactor environment. In the present paper the general physical mechanisms of radiation swelling of SiC are investigated. Recent experimental results concerning the effect of neutron and charged particle irradiation on radiation swelling of SiC are presented. A new theoretical model is suggested for the description of radiation swelling in ceramic materials. Point defects in ceramic materials can have an effective charge (e.g., an F+ center, vacancy with a single trapped electron). The theoretical model is based on kinetic consideration of charged point defect accumulation and kinetic growth of dislocation loops in the matrix taking into account the effect of internal electric field formed under irradiation in the matrix on diffusion processes of charged point defects. The theoretical results for radiation swelling are compared with the existing experimental data for irradiated SiC material.

Status of activities on the lithium target in the key element technology phase in IFMIF

This paper describes the status of the liquid lithium (Li) target facility in the key element technology phase (KEP) of the International Fusion Materials Irradiation Facility (IFMIF). The IFMIF is being jointly developed by the European Union (EU), Japan (JA), Russian Federation (RF) and US to provide an accelerator-based D–Li neutron source for testing the candidate materials for fusion reactors. A key issue of the Li target is to obtain a stable liquid Li flow with a speed of 20 m/s under a deuterium beam deposition of 10 MW. In the KEP, 19 tasks for the Li target are proposed and shared by EU, JA and RF. These tasks are a Li simulation experiment by water jet, Li flow experiment, corrosion/erosion, remote handling of the target assembly, and safety analysis. In addition to the KEP tasks, detailed design of the target is being performed.

Formula omitted]-Loop characterization in α-Fe: comparison between experiments and modeling

Ferritic/martensitic steels considered as first-wall candidate materials for fusion reactors experience significant radiation hardening at temperatures below ∼400 °C. Experimental evidence suggests that the observed defects are large interstitial dislocation loops with Burgers vector b=〈1 0 0〉 and b= 1 2 〈1 1 1〉 . In this work, the atomic character of 〈1 0 0〉 and 1 2 〈1 1 1〉 loops is investigated with molecular dynamics simulations and the atomic configurations are used to calculate the defect image contrast through direct simulation of TEM images. The simulated images are subsequently compared with actual TEM micrographs of irradiated ferritic materials and are in very good qualitative agreement, providing strong indication that the observed b=〈1 0 0〉 loops are interstitial in nature.